NUREG/CR-5625 (ORNL-6698) - Oak Ridge National Laboratory

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Procedure 5 PROPOSED REGULATORY GUIDE PROCEDURE This section of the report presents the proposed procedure for the initially loaded or reloaded reactor to the time at for use in a revision to the NRC regulatory guide on spent which the next reloaded core becomes critical. The fuel heat generation in an independent spent fuel storage exception is for the last cycle where the cycle ends with the installation. Section 5.1 contains the definitions, as used last reactor shutdown before discharge of the assembly. Ti here, of parameters needed in the determination of the denotes the elapsed time during cycle i for the assembly. heat generation rate of a fuel assembly. Section 5.2.1 Specifically, the first and last cycles are denoted by i = s (for contains the procedure for interpolating tables to derive start) and i = e (for end), respectively. T res, the total the uncorrected heat rate of an assembly. Sections 5.2.2- residence time of the assembly, is the sum of all T i for i = s 5.2.6 include the final evaluation method that uses simple through e, inclusive. Except for the last cycle for an adjustment factors for cases that are somewhat nontypical, assembly, the cycle times include the downtimes during in addition to the specified safety factor. reload. Cycle times, in this guide, are in days. There may be fuel assemblies with characteristics that lie sufficiently outside the mainstream of typical plant operations as to require a separate method for predicting the heat generation rate. Assemblies whose parameters lie outside the range of values used in the guide may be considered atypical for the purposes of using the proposed guide revision. A discussion of the characteristics of assumed typical reactor operations is in Sect. 5.3. A glossary of terms is given in Sect. 5.4. 5.1 Definitions and Derivations of Required Parameters The following definitions are used in the proposed guide procedure. 5.1.1 Heat Generation Rate of the Assembly (p) The heat generation rate of the spent fuel assembly is the recoverable thermal energy (from radioactive decay) of the assembly per unit time per unit fuel mass. The units for heat generation rate used in this guide are W per kg U, where U is the initial uranium loaded. Heat generation rate has also been referred to as decay heat rate, afterheat, or afterheat power. 5.1.2 Cycle and Cycle Times of the Assembly (T ) i A cycle of the operating history for a fuel assembly is the duration between the time criticality is obtained 5.1.3 Fuel Burnup of the Assembly (B i and B tot ) The fuel burnup of cycle i, B i , is the recoverable thermal energy per unit fuel mass during the cycle in units of megawatt days per metric ton (tonne) initial uranium * (MWd/tU) or in the SI units of mass used in the guide, megawatt day per kilogram U (MWd/kgU). B i is the maximum estimate of the fuel assembly burnup during cycle i. B tot is the total operating history burnup: B tot ' j e i's B i . 5.1.4 Specific Power of the Fuel (P i , P e , and P ave) (1) Specific power has a unique meaning in the guide. The reason for developing this definition is to take into account the differences between the actual operating history of the assembly and that used in the computation of the tabulated heat generation rates. The calculational model applied an uptime (time at power) of 80% of the cycle time in all except the last cycle (of the discharged fuel assembly), which had no downtime. The definition of specific power, used here, has two basic characteristics. First, when the actual uptime experienced by the assembly exceeds the 80% applied in the SAS2H/ORIGEN-S calculations, the heat rate changes by less than 1%. Second, when the actual uptime experienced is lower than the 80% applied in the calculations, the heat rate is reduced. * The adopted International System of units. 17 NUREG/CR-5625
Page 1 and 2: NUREG/CR-5625 ORNL-6698 Computing A
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Page 5 and 6: CONTENTS (continued) Page 6.6.3 Err
Page 7 and 8: LIST OF TABLES Page 2.1 List of nuc
Page 9 and 10: LIST OF TABLES (continued) Page D.1
Page 11 and 12: NUREG/CR-5625 xii
Page 13 and 14: 1 INTRODUCTION Heat is generated du
Page 15 and 16: 2 SOURCES OF DATA FOR COMPUTING HEA
Page 17 and 18: Comparisons Table 3.1 Point Beach U
Page 19 and 20: Comparisons Table 3.4 Turkey Point
Page 21 and 22: Comparisons Table 3.7 Cooper Nuclea
Page 23 and 24: Comparisons a Table 3.11 Element co
Page 25 and 26: Comparisons a Table 3.14 Turkey Poi
Page 27: Comparisons 4 HEAT RATE DATA COMPUT
Page 31 and 32: Procedure Table 5.1 BWR spent fuel
Page 33 and 34: Procedure Table 5.3. BWR spent fuel
Page 35 and 36: Procedure Table 5.6. PWR spent fuel
Page 37 and 38: Procedure Compute p tab , the heat
Page 39 and 40: Procedure Table 5.9 Enrichment fact
Page 41 and 42: 6 DISCUSSION OF THE PROPOSED PROCED
Page 43 and 44: Table 6.1 Comparison of heat rates
Page 45 and 46: Discussion Figure 6.2 Cooper Nuclea
Page 47 and 48: Discussion to sufficiently cover th
Page 49 and 50: Discussion 37 NUREG/CR-5625
Page 51 and 52: Discussion 39 NUREG/CR-5625
Page 53 and 54: Table 6.4. Excess power adjustment
Page 55 and 56: Discussion A large quantity of data
Page 57 and 58: F y y ' 1 y 3 j i'1 (f i y i F yi /
Page 59 and 60: Discussion Table 6.8 Percentage sta
Page 61 and 62: Discussion Table 6.10 Computed stan
Page 63 and 64: Discussion Table 6.11 Percentage st
Page 65 and 66: Discussion Table 6.13 Summary of cr
Page 67 and 68: This section deals with inaccuracie
Page 69 and 70: Discussion Table 6.16 Summary of pe
Page 71 and 72: 7 THE LWRARC CODE The LWRARC (Light
Page 73 and 74: LWRARC Code 3. Fuel type - Select f
Page 75 and 76: LWRARC Code Table 7.2 PWR fuel asse
Page 77 and 78: 8 SUMMARY Proper methods of storing
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9 REFERENCES 1. O. W. Hermann and R
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References 34. F. J. Sweeney and J.
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APPENDIX A DATA AND SAMPLE INPUT TO
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Appendix A Table A.3 Operating hist
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Appendix A ' ' ASSEMBLY AND CYCLE P
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Appendix A The following is the ent
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APPENDIX B SAMPLE CASE USING HEAT G
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APPENDIX C LWRARC CODE SAMPLE RESUL
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Appendix C ************************
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Appendix C ************************
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APPENDIX D ACTINIDE, FISSION PRODUC
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Appendix D 89 NUREG/CR-5625
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APPENDIX E PLOTS OF MAJOR DECAY HEA
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Appendix E 99 NUREG/CR-5625
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NUREG/CR-5625 ORNL-6698 INTERNAL DI
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107. D. A. Thomas, B&W Fuel Co., 10
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